Affiliation:
1. Harbin Engineering University
2. Heilongjiang Academy of Sciences
Abstract
In this work, molecular dynamics (MD) simulation was utilized in relation to access the thermal conductivity of UO2, PuO2 and (U, Pu)O2 in temperature range of 500–3000 K. Diffusion study on mixed oxide (MOX) was also performed to assess the effect of radiation damage by heavy ions at burnup temperatures. Analysis of the lattice thermal conductivity of irradiated MOX to its microstructure was carried out to enhance the irradiation defects with how high burnup hinders fuel properties and its pellet-cladding interaction. Fission gas diffusion as determined was mainly modelled by main diffusion coefficient. Degradation of diffusivity is predicted in MOX as composition deviate from the pure end members. The concentration of residual anion defects is considerably higher than that of cations in all oxides. Depending on the diffusion behavior of the fuel lattice, there was decrease in the ratio of anion to cation defects with increasing temperature. Besides, the modern mixed oxide fuel releases fission gas compared to that of UO2 fuel at moderate burnups.
Publisher
Trans Tech Publications, Ltd.
Subject
Condensed Matter Physics,General Materials Science,Atomic and Molecular Physics, and Optics
Reference21 articles.
1. Y. Guerin, G. S. Was, S. J. Zinkle, Materials Challenges for Advanced Nuclear Energy Systems. MRS Bulletin 34 (1), (2009) 10-14.
2. Basic Research Needs for Advanced Nuclear Energy Systems: Report of the Basic Energy Sciences Workshop on Basic Research Needs for Advanced Nuclear Energy Systems; U.S. Department of Energy Office of Basic Energy Sciences: (2006).
3. Simulation Based Engineering Science - Revolutionizing Engineering Science through Simulation; National Science Foundation: (2006).
4. G. S. Was, Fundamentals of Radiation Materials Science (Springer, Heidelberg, 2007).
5. M. S. Veshchunov, R. Dubourg, V. D. Ozrin, V. E. Shestak, and V. I. Tarasov. 2007. Mechanistic Modelling of Urania Fuel Evolution and Fission Product Migration during 320 Irradiation and Heating., Journal of Nuclear Materials. 321 https://doi.org/10.1016/j.jnucmat.2007.01.081.